Measurement data base on fluid mixing and flow distribution in the reactor circuit


Measurement data base on fluid mixing and flow distribution in the reactor circuit

Rohde, U.; Kliem, S.; Höhne, T.; Prasser, H.-M.; Hemström, B.; Toppila, T.; Elter, J.; Bezrukov, Y.; Scheuerer, M.

Experimental investigations on coolant mixing in Pressurised Water Reactors (PWR) have been performed within the EC project FLOMIX-R. The project was aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. Measurement data from a set of mixing experiments have been gained by using advanced measurement techniques with enhanced resolution in time and space. Slug mixing tests simulating the start-up of the first main circulation pump are performed with three 1:5 scaled facilities: the Rossendorf Coolant Mixing model ROCOM, the Vattenfall test facility and a metal mock-up of VVER-1000 type reactor at EDO Gidropress. Experimental results on buoyancy driven mixing of fluids with density differences have been obtained at ROCOM and the Fortum PTS test facility. In generic experiments with injection of water with higher density performed at ROCOM, transition between momentum driven mixing as it is typical for pump start-up scenarios, and buoyancy driven mixing was found. The Froude number was identified as a proper transition criterion. Measurement data available from NPP Paks VVER-440 type reactor commissioning tests together with data from the ROCOM facility are used as a basis for the flow distribution studies. Alltogether, a unique data base has been created to be used for the validation of Computational Fluid Dynamics (CFD) codes for the application to turbulent mixing in nuclear reactors.

Keywords: turbulent mixing; boron dilution; pre-stressed thermal shock; computational fluid dynamics; numerical simulation; measurement data base; pressurised water reactor

  • Lecture (Conference)
    The 11th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-11), 02.-06.10.2005, Avignon, France
  • Contribution to proceedings
    11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, 02.-06.10.2005, Avignon, France, 2-9516195-0-2

Permalink: https://www.hzdr.de/publications/Publ-7261
Publ.-Id: 7261